• Nuclear Engineering and Technology
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• 제39권3호
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• pp.161-170
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• 2007

Metal fuel, which was abandoned in the 1960's in favor of oxide fuel, has since then proven to be a viable fast reactor fuel. Key discoveries allowed high burnup capability and excellent steady-state as well as off-normal performance characteristics. Metal fuel is a key to achieving inherent passive safety characteristics and compact and economic fuel cycle closure based on electrorefining and injection-casting refabrication.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1747-1755
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• 2021

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

• Nuclear Engineering and Technology
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• 제35권4호
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• pp.309-317
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• 2003

The mass balance of the unit processes of the Advanced spent fuel Conditioning Process was calculated to obtain basic information. Based on this mass balance, the changes in decay heat and radioactivity of the spent fuel due to the metallization in the high temperature molten salt system were estimated. The decay heat and the radioactivity were calculated by using the ORIGEN2 computer code, and the result showed that the decay heat and the radioactivity of the metallized spent fuel ingot were $24.27\%\;and\;24.24\%$, respectively, compared to those of oxide spent fuel.

• Corrosion Science and Technology
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• 제14권6호
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• pp.301-312
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• 2015

Corrosion of zirconium fuel cladding is known to limit the lifetime and reloading cycles of fuel in nuclear reactors. Oxide layers formed on ZIRLO4^{TM}$ cladding samples, after immersion for 300-hour and 50-day in a simulated primary water chemistry condition ($360^{\circ}C$ and 20 MPa), were analyzed by using the scanning transmission electron microscopy (STEM), in-situ transmission electron microscopy (in-situ TEM) with the focused ion beam (FIB) technique, and X-ray diffraction (XRD). Both samples (immersion for 300 hours and 50 days) revealed the presence of the ZrO sub-oxide phase at the metal/oxide interface and columnar grains developed perpendicularly to the metal/oxide interface. Voids and micro-cracks were also detected near the water/oxide interface, while relatively large lateral cracks were found just above the less advanced metal/oxide interface. Equiaxed grains were mainly observed near the water/oxide interface.

• Nuclear Engineering and Technology
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• 제43권4호
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• pp.335-342
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• 2011

As part of the Department of Energy's Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

• Corrosion Science and Technology
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• 제13권4호
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• pp.145-151
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• 2014

The objectives of this study are to investigate the effect of the concentration ratios of Ni and Fe ions on crud deposition onto the fuel cladding surface in the simulated primary environments of a pressurized water reactor. Crud deposition tests were conducted in the Ni and Fe concentration ratios of 20:20 ppm, 39:1 ppm and 1:39 ppm at $325^{\circ}C$ for 14 days. In the case of the same Ni and Fe ion ratio (20:20), nickel ferrite with a polyhedral shape was formed. Nickel oxide deposits with a needle shape were formed in the condition of high Ni to Fe ion ratio (39:1), While polyhedral iron oxide and needle-like nickel oxide formed in the condition of low Ni to Fe ion ratio (1:39). The amount of deposits increased, when Fe oxides were formed. This indicates that Fe rich oxides stimulated Ni oxide deposition.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.304.2-304.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.65.1-65
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• 1999

• 공업화학
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• 제28권3호
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• pp.345-350
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• 2017

원자력연료 제조공정에서 생산되는 우라늄산화물(uranium oxide, UOX) 소결체의 밀도 분석은 일반적으로 소결공정을 거친 후, 소결체의 표본을 가지고 측정한다. 본 연구에서는 우라늄산화물의 중간물질인 중우라늄산암모늄(ammonium diuranate)의 색도를 분광기(spectrophotometer)로 측정함으로써 소결공정 이전에 우라늄산화물 소결체의 밀도를 분석해 보았다. 중우라늄산암모늄 표준 샘플 5개를 통해 얻은 명도 및 색의 좌푯(L, a, b)값과 통상적인 방법으로 얻은 소결체 밀도의 상관관계 추세선을 바탕으로 표적 샘플의 밀도를 분석한 결과, L 값에 대한 소결체의 밀도 분석이 결정계수 $R^2$ 값 0.9967로 가장 신뢰성이 높게 나왔음을 확인하였다. a 값에 대한 결정계수 $R^2$ 값은 0.9534로 상관관계가 높은 편이나 L 값보다는 낮았다. 이에 반해 b 값에 대한 결정계수 $R^2$ 값은 0.4349로 상관관계가 거의 없었다.

• Nuclear Engineering and Technology
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• 제35권5호
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• pp.471-481
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• 2003

In order to evaluate the microstructural characteristics of irradiated fuel claddings (Zircaloy-4), two irradiated specimens having different burnups (18 GWD/MTU and 42 GWD/MTU) were prepared from the G23-M4 fuel rods, which were loaded in Kori Unit 1 for 4 fuel cycles. The oxide thickness, hydride morphology and hardness change were characterized by an optical microscope and a micro-hardness tester after preparing the irradiated specimens in the PIE (Post Irradiation Examination) facilities. The dislocation loops and the amorphous transformation of precipitates induced by the neutron irradiation in the nuclear plant were also examined by a TEM. As the burnup increased from 18 GWD/MTU to 42 GWD/MTU, the oxide thickness increased from $6.0{\mu}m\;to\;25.3{\mu}m$ and the contents of hydrogen pick-up in the Zr matrix also greatly increased. In the comparison of the hardness of the unirradiated fuel cladding, the amount of hardness increase was nearly $9.3\%\;and\;24.2\%$ for the irradiated samples of 18 and 42 GWD/MTU, respectively. Both -type dislocation loops and -type dislocation components were observed simultaneously in the irradiated specimens and the densities of the dislocation was increased by increasing the burnup. The precipitates in both the irradiated specimens were amorphously transformed by the neutron irradiation and the trend of the amorphous transformation of the precipitates was enhanced at a higher burnup.